The present invention relates to heat exchange apparatus transferring heat from a reactor primary coolant, typically helium or carbon dioxide, to a secondary fluid medium, typically water and steam, and more particularly to a novel superheating arrangement in which a reheater tube bundle located within a nuclear pressure vessel works in conjunction with superheater tube bundles which are located outside of the nuclear pressure vessel. The reheater absorbs sufficient heat from the reactor gas coolant to supply required reheat steam to the reheat turbine, and in addition the reheater absorbs excess heat at higher temperature than required to meet the reheat turbine inlet steam conditions. The excess heat contained in the reheat steam flow is transferred regeneratively to the external superheater tube bundles to raise the superheat temperature of the main steam flow to the temperature required by the main steam turbine. While it is understood that various fluids can be used for the reactor primary coolant and the secondary fluid medium, the descriptions which follow shall employ the terminology reactor gas coolant to describe the reactor primary coolant and water and or steam to describe the secondary fluid medium.
It is desirable to remove heat from gas cooled nuclear reactors by circulating superheated steam at maximum temperature to maximize volumetric and thermal efficiency. This is typically done with tubular heat exchangers specifically referred to as steam generators. A steam generator is comprised of a series of high pressure main steam tube bundles which supply steam to the high pressure main steam turbine, and a lower pressure reheat tube bundle which supplies steam to the lower pressure reheat turbine. Within the nuclear pressure vessel the main steam tube bundle is comprised of an economizer/evaporator tube bundle stage in which feedwater is raised in temperature and evaporated to steam, and an initial superheater tube bundle stage in which the main steam flow is superheated to a desired level for exit from the nuclear pressure vessel. The intermediate superheater and the finishing superheater tube bundles are contained in separate pressure vessels located outside of the nuclear pressure vessel. Steam exiting from the initial superheater tube bundle stage is raised in temperature in the intermediate superheater tube bundle until stress limitations on the heat transfer tube material require a higher grade tube material. Accordingly, the finishing superheater tube bundle, pressure vessel and other components are constructed of materials having design stress limits high enough to accomodate the final steam temperature required by the main steam turbine. A bi-metallic weld is provided between the intermediate superheater tube bundle and the finishing superheater tube bundle. Inlet and outlet penetrations in the walls of the various pressure vessels provide for passage of water and steam flow to and from the respective tube bundles.
A steam generator can be designed to make steam at subcritical (less than 3206.2 psia) pressure or supercritical (greater than 3206.2 psia) pressure. In a subcritical system water changes to steam with heat addition at constant temperature and with water density exceeding steam density, while in a supercritical system the phase change is temperature dependant, occuring without a change in density. By employing a supercritical main steam system reheat steam pressure can be raised above reactor gas coolant pressure such that radiation bearing reactor gas coolant cannot leak into reheat steam.
Because of space limitations in the nuclear pressure vessel a once-through steam generator is preferred over a drum type steam generator in gas cooled reactors. However, once-through steam generators have certain inherent problems when utilized in gas cooled reactors. In prior designs utilizing once-through steam generators in gas cooled reactors parallel tube circuits were continuous from feedwater inlet to finishing superheater outlet so that steam temperature could not be equalized among tube circuits by the use of mixing headers. Also the lack of intermediate mixing headers and confinement in the nuclear pressure vessel precluded the use of water recirculation to provide flow stability (positive upward flow in all tube circuits) during low load and start-up operation of the plant. As a result flow resistance in the form of orifices at tube circuit inlets had to be provided. Orifices imposed a large pressure drop penalty and had a predisposition to foul by build-up of deposits from impurities in the feedwater. Special feedwater demineralizer systems had to be employed to reduce fouling of the otherwise non-maintainable orifices. Another problem with the use of once-through steam generators in gas cooled reactors was protection of the bi-metallic weld which had to be located within the nuclear pressure vessel in the tubing connecting the intermediate and finishing super-heater stages. Because the bi-metallic welds were not maintainable the use of special insulation and temperature sensors was required. It has been accepted practice with once-through steam generators in gas cooled reactors to plan for plugging of failed tubes because access for replacement of these tubes was not available. The potential for tube failure was high due to vibration and wear of tubes, blockage of tubes from orifice fouling, thermal stress at bi-metallic welds, and over heating due to low flow instability, poor gas and or water/steam flow distribution, and gas hot streaks and unmixed tube side flow.
The inability to provide recirculation flow during low load and start-up operation also limited main steam outlet pressure such that it was substantially lower than reactor gas coolant pressure. High safety gas cooled reactor designs eliminated reheating from the steam generator system because of potential leakage of radiation bearing reactor gas coolant into reheat steam, leading to further reduction of main steam outlet pressure. As a result the plant was deprived of several economic advantages including thermal and volumetric efficiency and the use of standard turbine equipment.
In general the difficulties with once-through steam generators and the lack of reheaters have prevented gas cooled reactors from realizing the very high temperature capability of the graphite core. The advantages of the present invention, namely a steam generator heat removal system having once-through capability at normal load combined with capability for water recirculation at low load and during start-up, working in conjunction with a balanced pressure reheater will avail gas cooled reactors of highest temperature potential.